Text

Technical Specification Pages for Amendments 210 and 204
	ML033580393
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Site:	Catawba
Issue date:	12/19/2003
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SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR for four loop operation; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained

> 1.14 for the WRB-2M CHF correlation.

2.1.1.2 The peak fuel centerline temperature shall be maintained < 5080 degrees F, decreasing 58 degrees F for every 10,000 MWd/mtU of fuel burnup.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained < 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

Catawba Units 1 and 2 2.0-1 Amendment Nos.210 & 204

Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and A0Os. The reactor core SLs are established to preclude violation of the following fuel design criteria:

a.

There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and

b.

The hot fuel pellet in the core must not experience centerline fuel melting.

The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, RCS flow rate, Al, pressure, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

Automatic enforcement of these reactor core SLs is provided by the appropriate operation of the RPS and the steam generator safety valves.

The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, 'RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," or the assumed initial conditions of the safety analyses (as indicated in the UFSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

SAFETY LIMITS The figure provided in the COLR shows the loci of points of fraction of RATED THERMAL POWER, RCS Pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, and that the exit quality is within the limits defined by the DNBR correlation.

The reactor core SLs are established to preclude violation of the following fuel design criteria:

a.

There must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criteria) that the hot fuel rod in the core does not experience DNB; and

b.

There must be at least a 95% probability at a 95% confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.

(continued)

Catawba Units 1 and 2 B 2.1.1-2 Revision No. 1

I Reactor Core SLs B 2.1.1 BASES SAFETY LIMITS (continued)

The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Overtemperature and Overpower AT reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and Al that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs.

APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT If SL 2.1.1 is violated, the requirement to go to MODE 3 places VIOLATIONS the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

REFERENCES

1.

10 CFR 50, Appendix A, GDC 10.

2.

UFSAR, Section 7.2.

I Catawba Units 1 and 2 B 2.1.1-3 Revision No. 1

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 7)

Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 4.3% (Unit 1) and 4.5% (Unit 2) of RTP.

AT (1 +

5) 1

) < ATO(l K, -K2 (1 + r s) [

1

+

T' 5

]

}

Where:

AT is the measured RCS AT by loop narrow range RTDs, 'F.

ATo is the indicated AT at RTP, "F.

s is the Laplace transform operator, sec1.

T is the measured RCS average temperature, OF.

T is the nominal Tavg at RTP (allowed by Safety Analysis), < the values specified in the COLR.

P is the measured pressurizer pressure, psig P is the nominal RCS operating pressure, = the value specified in the COLR K1

=

Overtemperature AT reactor NOMINAL TRIP SETPOINT, as presented in the COLR, K2

=

Overtemperature AT reactor trip heatup setpoint penalty coefficient, as presented in the COLR, K3

=

Overtemperature AT reactor trip depressurization setpoint penalty coefficient, as presented in the COLR,

-T, t 2

=

Time constants utilized in the lead-lag compensator for AT, as presented in the COLR, T3

=

Time constant utilized in the lag compensator for AT, as presented in the

COLR,

¶4, t5

=

Time.constants utilized in the lead-lag compensator for Tag, as presented in the COLR, T6

=

Time constant utilized in the measured Tavg lag compensator, as presented in the COLR, and f1(AI) =

a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for q, - qt between the "positive" and "negative" f1(Al) breakpoints as presented in the COLR; fl(Al) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent Al that the magnitude of q, - qb is more negative than the f1(AI) 'negative" breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the fI(Al) "negative" slope presented in the COLR; and (continued)

Catawba Units 1 and 2 3.3.1 -1 8 Amendment Nos. 210 & 204

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 7)

Reactor Trip System Instrumentation (iii) for each percent Al that the magnitude of qt - qb is more positive than the fl(Al) "positive" breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the f1(Al) "positive" slope presented in the COLR.

Note 2: Overpower AT The Overpower AT Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 2.6% (Unit 1) and 3.1% (Unit 2) of RTP.

AT (1 +

(1

)< AT. K-K5 T7 S 1

) T-K6 T 1 sT'1 f2 (A,)}

Where:

AT is the measured RCS AT by loop narrow range RTDs, OF.

ATo is the indicated AT at RTP, OF.

s is the Laplace transform operator, see.

T is the measured RCS average temperature, "F.

T is the nominal Tavg at RTP (calibration temperature for AT instrumentation),

< the values specified in the COLR.

K4

=

Overpower AT reactor NOMINAL TRIP SETPOINT as presented in the

COLR, K5

=

the value specified in the COLR for increasing average temperature and the value specified in the COLR for decreasing average temperature, K6

=

Overpower AT reactor trip heatup setpoint penalty coefficient as presented in the COLR for T > T and K6 = the value specified in the COLR for T < T',

t1,,r2

= Time constants utilized in the lead-lag compensator for AT, as presented in the COLR,

¶3

=

Time constant utilized in the lag compensator for AT, as presented in the

COLR,

'r6

=

Time constant utilized in the measured Tavg lag compensator, as presented in the COLR, T.

=

Time constant utilized in the rate-lag controller for Tavg, as presented in the COLR, and f2(AI) =

a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for qt - qb between the "positive" and 'negative" f2(AI) breakpoints as presented in the COLR; f2(AI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (continued)

Catawba Units I and 2 3.3.1 -1 9 Amendment Nos.210 & 204

RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) setpoints are based on the minimum flow specified in the COLR. Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE in MODE 1 above P-8.

In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core. In MODE 1 below the P-8 setpoint, a loss of flow in two or more loops is required to actuate a reactor trip (Function 1 O.b) because of the lower power level and the greater margin to the design limit DNBR.

b.

Reactor Coolant Flow-Low (Two Loops)

The Reactor Coolant Flow-Low (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in two or more RCS loops while avoiding reactor trips due to normal variations in loop flow.

Above the P-7 setpoint and below the P-8 setpoint, a loss of flow in two or more loops will initiate a reactor trip. The setpoints are based on the minimum flow specified in the COLR. Each loop has three flow detectors to monitor flow.

The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE.

In MODE 1 above the P-7 setpoint and below the P-8 setpoint, the Reactor Coolant Flow-Low (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on low flow are automatically blocked since power distributions that would cause a DNB concern at this low power level are unlikely. Above the P-7 setpoint, the reactor trip on low flow in two or more RCS loops is automatically enabled. Above the P-8 setpoint, a loss of flow in any one loop will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR.

Catawba Units 1 and 2 B 3.3.1-1 8 Revision No. 2

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 APPLICABILITY:

RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in Table 3.4.1-1.

MODE 1.

- -- - - -- - - - -- - - -- N r----------------m----------------------------------------

Pressurizer pressure limit does not apply during:

a.

THERMAL POWER ramp > 5% RTP per minute; or

b.

THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressurizer pressure or A.1 Restore DNB parameter(s) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months RCS average to within limit.

temperature DNB parameters not within limits.

B.

RCS total flow rate >

B.1 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 99%, but < 100% of the POWER to < 98% RTP.

limit specified in the COLR.

AND B.2 Reduce the Power Range 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Neutron Flux - High Trip Setpoint below the nominal setpoint by 2% RTP.

(continued)

Catawba Units 1 and 2 3.4.1 -1 Amendment Nos.210 & 204

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

RCS total flow rate <

C.1 Restore RCS total flow rate 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 99% of the value to > 99% of the value specified in the COLR.

specified in the COLR.

OR C.2.1 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER to < 50% RTP.

AND C.2.2 Reduce the Power Range 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Neutron Flux - High Trip Setpoint to < 55% RTP.

AND C.2.3 Restore RCS total flow rate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to > 99% of the value specified in the COLR.

D.

Required Action and D.1 Be in MODE 2.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

I Catawba Units 1 and 2 3.4.1 -2 Amendment Nos.210 & 204

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 Table 3.4.1-1 (page 1 of 1)

RCS DNB Parameters No. OPERABLE PARAMETER INDICATION CHANNELS LIMITS

1.

Indicated RCS Average meter 4

< the value specified Temperature - Unit 1 in the COLR meter 3

< the value specified in the COLR computer 4

<the value specified in the COLR computer 3

< the value specified in the COLR Indicated RCS Average meter 4

< the value specified Temperature - Unit 2 in the COLR meter 3

< the value specified In the COLR computer 4

< the value specified in the COLR computer 3

< the value specified in the COLR

2.

Indicated Pressurizer meter 4

> the value specified Pressure In the COLR meter 3

> the value specified in the COLR computer 4

> the value specified in the COLR computer 3

> the value specified in the COLR

3.

RCS Total Flow Rate

> 388,000 gpm and >

the limit specified in the COLR (Unit 1);

> 390 000 gpm and >

the limit specified In the COLR (Unit 2)

Catawba Units 1 and 2 3.4.1 -4 Amendment Nos.21 0 & 204

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits BASES I

BACKGROUND These Bases address requirements for maintaining RCS pressure, temperature, and flow rate within limits assumed in the safety analyses.

The safety analyses (Ref. 1) of normal operating conditions and anticipated operational occurrences assume initial conditions within the normal steady state envelope. The limits placed on RCS pressure, temperature, and flow rate ensure that the minimum departure from nucleate boiling ratio (DNBR) will be met for each of the transients analyzed.

The RCS pressure limit is consistent with operation within the nominal operational envelope. Pressurizer pressure indications are averaged to come up with a value for comparison to the limit. A lower pressure will cause the reactor core to approach DNB limits.

The RCS coolant average temperature limit is consistent with full power operation within the nominal operational envelope. Indications of temperature are averaged to determine a value for comparison to the limit.

A higher average temperature will cause the core to approach DNB limits.

The RCS volumetric flow rate normally remains constant during an operational fuel cycle with all pumps running. Flow rate indications are averaged within a loop and then summed among the four loops to come up with a value for comparison to the limit. A lower RCS flow will cause the core to approach DNB limits. RCS flow rate may be slightly reduced provided THERMAL POWER is also reduced to ensure that the calculated DNBR will not be below the design DNBR value.

Operation outside these DNB limits increases the likelihood of a fuel cladding failure in a DNB limited event.

APPLICABLE SAFETY ANALYSES The requirements of this LCO represent the initial condition for transients analyzed in the plant safety analyses (Ref. 1). The safety analyses have shown that transients initiated from the limits of this LCO will result in meeting the acceptance criteria, including the DNBR criterion. This is the acceptance limit for the RCS DNB parameters. Changes to the unit that could impact these parameters must be assessed for their impact on the acceptance criteria. A key assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.6, "Control Bank Insertion Limits"; LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD);"

and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)."

Catawba Units 1 and 2 B 3.4.1 -1 Revision No. 2

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued)

The pressurizer pressure limits and the RCS average temperature limits specified in the COLR correspond to analytical limits used in the safety analyses, with allowance for measurement uncertainty.

The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2).

LCO This LCO specifies limits on the monitored process variables-pressurizer pressure, RCS average temperature, and RCS total flow rate-to ensure the core operates within the limits assumed in the safety analyses. The numerical limits of these variables are contained in the COLR to provide operating and analysis flexibility from cycle to cycle. However, the minimum RCS flow, based on previously analyzed maximum steam generator tube plugging for Unit 1 and the original licensed value for Unit 2, is retained in the LCO. Operating within these limits will result in meeting the acceptance criteria, including the DNBR criterion.

RCS total flow rate contains a measurement error based on the performance of past precision heat balances and using the result to calibrate the RCS flow rate indicators. Sets of elbow tap coefficients, as determined during these heat balances, were averaged for each elbow tap to provide a single set of elbow tap coefficients for use in calculating RCS flow. This set of coefficients establishes the calibration of the RCS flow rate indicators and becomes the set of elbow tap coefficients used for RCS flow measurement. Potential fouling of the feedwater venturi, which might not have been detected, could have biased the result from these past precision heat balances in a nonconservative manner. Therefore, a penalty for undetected fouling of the feedwater venturi raises the nominal flow measurement allowance for no fouling.

The numerical values for pressure and average temperature specified in the COLR are given for the measurement location with adjustments for the indication instruments.

APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough that DNB is not a concern.

A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be Catawba Units 1 and 2 B 3.4.1-2 Revision No. 2

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued) counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.

The DNBR limit is provided in SL 2.1.1, "Reactor Core SLs." Those limits are less restrictive than the limits of this LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action. Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded.

ACTIONS A.1 Pressurizer pressure and RCS average temperature are controllable and measurable parameters. With one or both of these parameters not within LCO limits, action must be taken to restore parameter(s).

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.

B.1 and B.2 RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. If the indicated RCS total flow rate is

> 99%, but < 100% of the limit specified in the COLR, then THERMAL POWER may not exceed 98% RTP. THERMAL POWER must be reduced within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is consistent with Required Action A.1. In addition, the Power Range Neutron Flux -

High Trip Setpoint must be reduced from the nominal setpoint by 2% RTP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reset the trip setpoints recognizes that, with power reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System.

C.1, C.2.1, C.2.2, and C.2.3 If the indicated RCS total flow rate is < 99% of the value specified in the COLR, then RCS total flow must be restored to 2 99% of the value specified in the COLR within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or power must be reduced to less than 50% RTP. The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is consistent with Required Action A.1. If THERMAL POWER is reduced to less than 50%

Catawba Units 1 and 2 B, 3.4.1-3 Revision No. 2

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES ACTIONS (continued)

RTP, the Power Range Neutron Flux - High Trip Setpoint must also be reduced to < 55% RTP. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reset the trip setpoints is consistent with Required Action B.2. This is a sensitive operation that may inadvertently trip the Reactor Protection System.

Operation is permitted to continue provided the RCS total flow is restored to 2 99% of the value specified in the COLR within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is reasonable considering the increased margin to DNB at power levels below 50% and the fact that power increases associated with a transient are limited by the reduced trip setpoint.

D.1 If the Required Actions are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable to reach the required plant conditions in an orderly manner.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS pressure and related equipment status. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

SR 3.4.1.2 This surveillance demonstrates that the average RCS temperature remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

Catawba Units 1 and 2 B 3.4.1-4 Revision No. 2

Reporting Requirements Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1.

Illustration of Reactor Core Safety Limits for Specification 2.1.1,

2.

Moderator Temperature Coefficient BOL and EOL limits and 60 ppm and 300 ppm surveillance limits for Specification 3.1.3,

3.

Shutdown Bank Insertion Limit for Specification 3.1.5, l

4.

Control Bank Insertion Limits for Specification 3.1.6, l

5.

Axial Flux Difference limits for Specification 3.2.3, l

6.

Heat Flux Hot Channel Factor for Specification 3.2.1,

7.

Nuclear Enthalpy Rise Hot Channel Factor for Specification 3.2.2, l

8.

Overtemperature and Overpower Delta T setpoint parameter I

values for Specification 3.3.1,

9.

Reactor Coolant System Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits for Specification 3.4.1,

10.

Accumulator and Refueling Water Storage Tank boron l

concentration limits for Specification 3.5.1 and 3.5.4,

11.

Reactor Coolant Systerri and refueling canal boron concentration I

limits for Specification 3.9.1,

12.

Spent fuel pool boron concentration limits for Specification 3.7.15, l

13.

SHUTDOWN MARGIN for Specification 3.1.1, l

14.

31 EFPD Surveillance Penalty Factors for Specifications 3.2.1 and l 3.2.2, and

15.

Reactor Makeup Water Pumps Combined Flow Rates limit for l

Specifications 3.3.9 and 3.9.2.

(continued)

Catawba Units 1 and 2 5.6-3 Amendment Nos. 210 & 204

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY" (W Proprietary).

2.

WCAP-10266-P-A, 'THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE" (W Proprietary).

3.

BAW-1 01 68-P-A, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants" (B&W Proprietary).

4.

DPC-NE-201 1-P-A, "Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors" (DPC Proprietary).

5.

DPC-NE-3001-P-A, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology" (DPC Proprietary).

6.

DPC-NF-2010-A, "Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design."

7.

DPC-NE-3002-A, "FSAR Chapter 15 System Transient Analysis Methodology."

8.

DPC-NE-3000-P-A, "Thermal-Hydraulic Transient Analysis Methodology' (DPC Proprietary).

9.

DPC-NE-1004-A, "Design Methodology Using CASMO-3/SIMULATE-3P."

10.

DPC-NE-2004-P-A, "Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01" (DPC Proprietary).

11.

DPC-NE-2005-P-A, "Thermal Hydraulic Statistical Core Design Methodology" (DPC Proprietary).

(continued)

Catawba Units 1 and 2 5.6-4 Amendment Nos. 210 & 204

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

12.

DPC-NE-2008-P-A, "Fuel Mechanical Reload Analysis Methodology Using TACO3" (DPC Proprietary).

13.

WCAP-10054-P-A, 'Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code" (W Proprietary).

14.

DPC-NE-2009-P-A, 'Westinghouse Fuel Transition Report" (DPC Proprietary).

15.

WCAP-12945-P-A, Volume 1 and Volumes 2-5, "Code Qualification Document for Best-Estimate Loss of Coolant Analysis" (W Proprietary).

The COLR will contain the complete identification for each of the Technical Specifications referenced topical reports used to prepare the COLR (i.e., report number, title, revision number, report date or NRC SER date, and any supplements).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Ventilation Systems Heater Report When a report is required by LCO 3.6.10, "Annulus Ventilation System (AVS),"

LCO 3.7.10, 'Control Room Area Ventilation System (CRAVS)," LCO 3.7.12, Auxiliary Building Filtered Ventilation Exhaust System (ABFVES)," LCO 3.7.13, "Fuel Handling Ventilation Exhaust System (FHVES)," or LCO 3.9.3, "Containment Penetrations," a report shall be submitted within the following 30 days. The report shall outline the reason for the inoperability and the planned actions to return the systems to OPERABLE status.

(continued)

Catawba Units I and 2 5.6-5 Amendment Nos. 210 & 204

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.7 PAM Report When a report is required by LCO 3.3.3, "Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.8 Steam Generator Tube Inspection Report

a.

The number of tubes plugged in each steam generator shall be reported to the NRC within 15 days following completion of the program;

b.

The complete results of the Steam Generator Tube Surveillance Program shall be reported to the NRC within 12 months following the completion of the program and shall include:

1.

Number and extent of tubes inspected,

2.

Location and percent of wall-thickness penetration for each indication of an imperfection, and

3.

Identification of tubes plugged.

c.

The results of inspections of steam generator tubes which fall into Category C-3 shall be reported to the NRC within 30 days prior to the restart of the unit following the inspection. This report shall provide a description of the tube degradation and corrective measures taken to prevent recurrence.

Catawba Units 1 and 2 5.6-6 Amendment Nos.210 & 204

v t e Letter Sequence Other
TAC:MB8359, Relocate TS Parameters to Colr (Open) TAC:MB8360, Relocate TS Parameters to Colr (Open)
Initiation Request, Request Acceptance... Administration Questions 2 October 2003 Answers Results Approval... Other: ML033140352, ML033570127, ML033580393
MONTHYEAR2003-11-06 [Table View]

ML033580393

Text

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